This invention relates to a method for immobilizing radioactive wastes for permanent disposal. More particularly, the invention relates to a method of immobilizing mixed waste chloride salts containing radionuclides and other hazardous materials for permanent disposal.
The recovery of fissionable materials such as uranium and plutonium from spent nuclear reactor fuels can be carried out by electrorefining methods using electrochemical cells of the type described in U.S. Pat. Nos. 4,596,647 and 2,951,793, as well as U.S. Pat. No. 4,880,506. It is the electrorefining method which is being developed for the reprocessing of Integral Fast Reactor (IFR) fuel. In a typical electrorefining cell an electrolyte consisting of a molten eutectic salt mixture such as KCl and LiCl is used to transport the metal or metals to be purified between electrode solutions. When used to reprocess spent nuclear reactor fuels, the salt mixture becomes contaminated with radionuclides, such as .sup.137 cesium, .sup.90 strontium and .sup.129 iodine, hazardous materials such as barium and other species such as sodium, and eventually is no longer suitable for use in the electrorefining cell.
Ideally, the salt would be decontaminated by removing a fraction of the heat-producing radionuclides, primarily cesium and strontium, and any other elements e.g. barium and sodium, which could potentially interfere in the operation of the electrorefiner, and the purified salt would be recycled back to the electrorefiner. However, the separation of cesium and strontium from the salt is difficult, and if they are separated in concentrated form, it would be necessary to dilute them in another matrix material and/or cool them before they could be stored since they are large heat producers. It is, therefore, more practical to dispose of the cesium and strontium and any other radionuclides, including iodides, and toxic metal chlorides along with a portion of the salt matrix. The waste salt containing the cesium, strontium and iodine is a high level waste (HLW), and as such must be disposed of in the geologic repository for HLW. This requires that the waste form be leach resistant to prevent an uncontrolled release of the radionuclides and other hazardous chemicals such as barium, into the groundwater. Since waste salts are chlorides and are very water soluble, a method for encapsulating and immobilizing the waste salt must be identified.
One problem with developing a waste storage medium is that the waste salt consists primarily of chloride salts of alkali metals and as such is not readily amenable to treatment using procedures and techniques developed for immobilizing cesium and strontium in other nuclear waste streams. For instance, the chloride salts cannot be added directly to glass-forming compounds and processed to yield a leach-resistant glass since glasses containing halide ions are relatively water soluble. Therefore, for immobilization in a glass matrix the waste chloride salts must be converted into oxides or other chemical forms compatible with the glass-making process. However, conversion processes are expensive and time-consuming and raise environmental concerns about off-gases produced by the processes. A mortar matrix has also been considered as a possible waste form for the waste chloride salt. A special mortar was developed to incorporate lithium, potassium, cesium and strontium chloride salts into its structure, thereby immobilizing them. However, when irradiated, the water in the mortar was radiolyzed and hydrogen gas was generated.
U.S. patent application Ser. No. 744,753, filed Aug. 14, 1991, and incorporated herein by reference, describes the use of certain zeolites to decontaminate and immobilize waste chloride salts. Contact between the zeolite (for example, zeolite A or mixtures of chabazite and erionite zeolites or mixtures thereof) in the sodium, potassium, or lithium form and the molten salt result in an ion exchange between the radionuclides cesium and strontium and the hazardous material barium in the salt and the sodium, potassium or lithium in the zeolite and the occlusion of up to about 25 wt % of the salt within the molecular cavities of the zeolite. This method has the advantage over many methods in that the radionuclides and barium are concentrated in the zeolite so that some of the salt partially purified of cesium, strontium and barium might be recycled back to the electrorefiner. Although this method is effective for purifying the salt, the method requires the removal of the non-occluded surface salt from the zeolite before it can be sent to storage. Furthermore, problems have been encountered in making dense, leach-resistant waste forms directly from the salt-occluded waste product.
The use of synthetic naturally occurring minerals to store radioactive ions has also been studied. U.S. Pat. No. 4,808,318 describes the use of a modified phlogopite to recover cesium ions from waste solutions. The modified phlogopite containing the cesium ion is then fixed so that it can be safely stored for a long period of time. U.S. Pat. No. 4,229,317 describes a method whereby radioactive iodine, present as alkali metal iodides or iodates is incorporated into a solid by adding appropriate amounts of alkali metal, alumina and silica to the solution, stirring to form a homogenous mixture, drying the mixture to form a powder and compacting and heating the powder under conditions appropriate to form sodalite, whereby the iodine ion is incorporated within the molecular cage of the sodalite (Na.sub.6 [(SiO.sub.2).sub.6 (AlO.sub.2).sub.6 ]2NaCl).
What is still needed is a method of immobilizing mixtures of salts, particularly chloride salts containing radionuclides and other hazardous wastes, so that the highly soluble salts can be safely stored for long periods of time in HLW storage facilities without presenting a hazard to the environment.